The use of an organic coolant and neutron moderator in a thermal nuclear reactor has already been proposed, and in particular the United States Organic Moderator Reactor Experiment (OMRE) was designed to use organic liquid for both coolant and moderator purposes but in separate circuits. The reactor utilised a separate core vessel containing relatively low temperature organic liquid moderator, with a relatively slow circulation rate. The organic coolant was passed over the fuel in the fuel channel tubes at a substantially higher temperature. However after only a short period of operation the concept was abandoned due to the high degradation rates of the organic oolant in the moderator circuit and on occasion in the coolant circuit.
A development of this experimental reactor was a power demonstration reactor at Piqua, Ohio. This reactor utilised a single pass of coolant, but the moderator region and fuel regions were separated with different flow velocities. This feature gave rise to unacceptable degradation of the coolant and the reactor was shut down after the initial period of operation. Piqua is the only example of an organic liquid moderated and cooled power reactor to proceed to construction and operation even though its limited operation (1964 to 1966) was unsuccessful and did not establish the feasibility of organic coolant.
With the advantage of that experience and as part of the Canadian deuterium-uranium (CANDU) programme, organic liquid coolant had been recognised as potentially advantageous and the Canadian Organic Cooled Reactor (CANDU-OCR) was developed, which used organic liquid as coolant only, the moderator being heavy water. A prototype reactor, the Whiteshell WR-1, has been built and its successful operation has demonstrated the technology and use of organic coolant. Commerical designs have been proposed but have not been implemented.
A major consideration in all nuclear reactor design, whatever the nature of the coolant and the moderator, is safety. Designers are constantly striving towards a so-called `inherently safe` reactor, by which is meant one whose safeguard systems are of such a design that upon the occurrence of fault conditions up to and including the most severe design postulated fault (sometimes termed the largest credible accident) they maintain the reactor in a safe condition for an extended period without any corrective action by the operators, before further safeguard actions are required. The length of that extended period is not defined, but it should preferably be a minimum of several hours and preferably up to a week. It is of course up to the designer to identify and specify the most severe design postulated fault to which any particular reactor has been designed. To date there are no known proposals for an inherently safe reactor that will withstand a loss of the whole of the normal operating inventory of both coolant and moderator.